Outgassing measurement of the waveguide module for a steady state LHCD antenna

An outgassing rate is measured using a waveguide module during a long pulse r.f. injection at 3.7 GHz in order to operate a lower hybrid current drive (LHCD) antenna in steady state. The waveguide module consists of four sub-waveguides which are made of dispersion-strengthened copper and copper-plated stainless steel with cooling channels to control its working temperature between 100 and 500 °C. The waveguide module shows the high power capability of power density up to 200 MW m⁻² after short term conditioning. A low outgassing rate of about 2 × 10⁻⁷ Pa m³ s⁻² m⁻² is obtained during r.f. injection up to 150 MW m⁻² at a working temperature of T=300 °C after 450 °C baking. Long pulse r.f. injection is effective for reduction of the outgassing by about 1/100. Outgassing during r.f. injection depends on the working temperature but is independent of the r.f. power density after sufficient conditioning. The calculation code taking desorption, adsorption and diffusion processes into account can show time behavior of outgassing in the waveguide module. A quasi-constant outgassing of about 10⁻⁷ Pa m³ s⁻¹ m⁻² is observed at a power density of 150 MW m⁻² in 1800 s injection while keeping a saturated temperature at the center of the module below 120 °C by using water cooling. The outgassing properties obtained indicate that steady state operation of an LHCD antenna is feasible without a large pumping system.     

De-agglomeration of thorium oxalate - a method for the synthesis of sinteractive thoria

Thorium oxalate was obtained by precipitation in water and in non-aqueous solvents and de-agglomerated by ultrasonication in both aqueous as well as non-aqueous media. Sinteractive thoria (crystallite size 6-20 nm) obtained from the de-agglomerated thorium oxalate was characterised for residual carbon, crystallite size, specific surface area, particle size distribution and bulk density. Microstructure of the precursor and the product was studied using TEM and HRTEM. The morphology of the sintered pellets was studied using SEM. The reactivity of the calcined powders was determined by measuring the density of the sintered compacts. The solvent used for de-agglomeration was found to have significant influence on the microstructure of the powders. Thoria derived through aqueous precipitation route could be sintered to a density of 9.7 Mg m⁻³ at 1673 K. It was demonstrated that ultrasonic de-agglomeration could be a useful method for obtaining sinteractive thoria.     

Study of radiation-induced degradation of RPV steels and model alloys by positron annihilation and Mössbauer spectroscopy

The influence of different microstructural processes on the degradation due to radiation embrittlement has studied by positron annihilation and Mössbauer spectroscopy. The materials studied consisted of WWER-440 base (15Kh2MFA) and weld (10KhMFT) RPV steels which were neutron-irradiated at fluence levels of 0.78 × 10²⁴ m⁻², 1.47 × 10²⁴ m⁻² and 2.54 × 10²⁴ m⁻²; WWER-1000 base (15Kh2NMFAA) and weld (12Kh2N2MAA) irradiated at a fluence level 1.12 × 10²⁴ m⁻²; three different model alloys implanted with protons at two dose levels (up to 0.026 dpa), finally the base metal of WWER-1000 (15Kh2NMFAA) was thermally treated with the intention to simulate the P-segregation process. It has been shown possible to correlate the values of parameters obtained by such techniques and data of mechanical testing (ductile-to-brittle transition temperature and upper shelf energy).     

Development of ceramic waste forms for actinide-rich waste— Radiation stability of perovskite and phase and chemical stabilities of ZR-and AL-based ceramics

Ceramics are considered as most promising materials for conditioning of long-lived radionuclides because of their outstanding durability for long term. The Japan Atomic Energy Research Institute (JAERI) has developed ceramic waste forms, e.g. Synroc and zirconia-based ceramics, for the actinide-rich wastes arising from partitioning and transmutation processes. In the present study, -decay damage effects on the density and leaching behavior of perovskite (one of three main minerals forming Synroc) were investigated by an accelerated experiment using the actinide doping technique. A decrease in density of Cm-doped perovskite reached 1.3 % at a dose of 9 × 10¹⁷ -decays·g⁻¹. The leach rates (MCC-1 leach test inpH 2 solution at 90°C for 2 months) of perovskite specimens with accumulated doses of 1.6 × 10¹⁷, 4.0 × 10¹⁷ and 8.3 × 10¹⁷ -decays·g⁻¹ were 1.7, 2.3 and 3.0 μ·m⁻²·day⁻¹, respectively. Application of zirconia- and alumina-based ceramics for incorporating actinides was also investigated by the experiments using non-radioactive elements (Ce and Nd) with an emphasis on crystallographic phase stability and chemical durability. The yttria-stabilized zirconia was stable crystallographically in the wide ranges of Ce and/ or Nd content and had excellent chemical durability.     

Effects of xenon implantation in spinel-zirconia/ceria composites

Two prethinned spinel specimens containing either Y_0.15Zr_0.85O₂ or Ce_0.5Zr_0.5O₂ particles were implanted with 200–400 keV Xe ions at 873 K using the IVEM-Tandem Facility at Argonne National Laboratory. In situ transmission electron microscopy (TEM) was conducted during the implantation in order to follow the evolution of the microstructure. At an ion fluence between 2.4x10²⁰ to 3x10²⁰ m⁻² (up to 50 dpa and 4.7 at %), large Xe bubbles of 50–100 nm developed at the boundaries of the small oxide particles, while a high density of dislocation loops (up to 8 nm in diameter) and much smaller bubbles (up to 4 nm in diameter) formed in the spinel matrix. No large bubbles were observed at the boundaries between the spinel grains. These results suggest that the boundaries between spinel and oxide particles are preferred sites for fission gas accumulation.     

Post-irradiation examination of uranium-doped rock-like oxide fuels

Post-irradiation examinations of rock-like oxide fuels were performed in JAERI to evaluate irradiation behavior and geochemical stability. Five kinds of fuels were prepared using 20% enriched U instead of Pu; a single-phase fuel of an yttria-stabilized zirconia containing UO₂ (U-YSZ), two particle-dispersed type fuels of U-YSZ particles + MgAl₂O₄/Al₂O₃ powder, two homogeneously blended type fuels of U-YSZ powder + MgAl₂O₄/Al₂O₃ powder. The fuels were irradiated in JRR-3 for about 100 days and estimated irradiation conditions were as follows; linear power was 15 kW m⁻¹, thermal neutron fluence was 7 x 10²⁴ m⁻² and fuel temperatures at the surface were 740–1130 K. From the results of non-destructive examinations, the stainless steel cladding surfaces were partially discolored by oxidation and no remarkable deformation of the pins was observed. Significant pellet fragmentation was not observed in spite of the crack formation as observed in irradiated LWR UO₂ fuels. Nonvolatile FPs were observed only in pellets but volatile Cs moved partly to the plenum. From these examinations, no significant difference in macroscopic irradiation behavior was distinguished among 5 fuels.     

Rock-like oxide fuel behavior under reactivity initiated accident conditions

Pulse irradiation tests of two types of rock-like oxide (ROX) fuel, i.e. yttria stabilized zirconia (YSZ) and YSZ/Spinel composite, were conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under reactivity-initiated accident conditions. The ROX fuels failed with cladding burst at fuel volumetric enthalpies above 10 GJ m⁻³, which was comparable to that of UO₂ fuel. The failure of the ROX fuels, however, occurred with considerable fuel melting and was quite different to that of UO₂ fuel, which was caused by cladding melting and embrittlement due to heavy oxidation. Lower fuel melting temperature of the ROX fuels compared to that of UO₂ contributed to the different fuel failure modes. Certain amount of molten ROX fuel dispersed out at the failure. However, the mechanical energy generation due to the molten fuel/water interaction was negligible for the ROX fuels at peak fuel enthalpies below 12 GJ m⁻³.     

Density measurement of amorphous SixGe1 − x alloys

The atomic density of amorphous Si_xGe_{1 − x} alloys (x = 1, 0.85, 0.67, 0.50, 0.20 and 0) has been measured. Mono-crystalline Si_xGe_1−x layers Were implanted with 1.50–2.75 MeV Si²⁺ and Ge²⁺ ions to produce the amorphous material. Using surface profilometry and RBS/channeling, it was found that amorphous alloys are less dense than the crystalline alloys, and that Vegard's law underestimates the a-Si_xGe_1−x density.     

Influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in PWR primary water

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670 °C and 720 °C for durations ranging from 5 h to 100 h. Spherical γ′ phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 × 10²² m⁻³ to 2 × 10²³ m⁻³ were measured. Results suggest that both the γ′ phase mean diameter and density quickly saturate with time for ageing heat treatment at 720 °C while the γ′ mean diameter increases significantly up to 100 h for ageing heat treatment at 670 °C. Grain boundary η phase precipitates were systematically observed for ageing heat treatment at 720 °C even for short ageing periods. In contrast, no grain boundary η phase precipitates were observed for ageing heat treatments at 670 °C except after 100 h. Hardening by γ′ precipitation was well described by the dispersed barrier hardening model with a γ′ barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 × 10⁻⁷ s⁻¹ in simulated PWR primary water at 320 °C and 360 °C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary η phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.     

Lattice location and electrical conductivity in ion implanted TiO2 single crystals

Single crystals of TiO₂ (rutile) were implanted at room temperature with Ar, Sn and W ions applying fluences of 10¹⁵/cm² to 10¹⁶/cm² at 300 keV. The lattice location, together with ion range and damage distribution was measured with Rutherford Backscattering and Channeling (RBS-C). The conductivity, σ, was measured as a function of temperature. The implanted Sn and W atoms were entirely substitutional on Ti sites in the applied fluence region, where the radiation damage did not yet reach the random level. A large σ increase was observed for all implants at displacement per atom values (dpa) below 1. Above dpa = 1, σ reveals a saturation value of 0.3 Ω⁻¹ cm⁻¹ for Ar implants, while for W and Sn implants a further increase of σ up to 30 Ω⁻¹ cm⁻¹ was measured. Between 70 K and 293 K ln σ was proportional to T^−1/2, (Ar,W) and T^−1/4 (Sn), indicating that the transport mechanism is due to variable range hopping.     

Thermodynamic studies on Cs4U5O17(s) and Cs2U2O7(s) by emf and calorimetric measurements

The oxygen potentials over the phase field: Cs₄U₅O₁₇(s)+Cs₂U₂O₇(s)+Cs₂U₄O₁₂(s) was determined by measuring the emf values between 1048 and 1206 K using a solid oxide electrolyte galvanic cell. The oxygen potential existing over the phase field for a given temperature can be represented by: Δμ(O₂) (kJ/mol) (±0.5)=−272.0+0.207T (K). The differential thermal analysis showed that Cs₄U₅O₁₇(s) is stable in air up to 1273 K. The molar Gibbs energy formation of Cs₄U₅O₁₇(s) was calculated from the above oxygen potentials and can be given by, Δ_fG⁰ (kJ/mol)±6=−7729+1.681T (K). The enthalpy measurements on Cs₄U₅O₁₇(s) and Cs₂U₂O₇(s) were carried out from 368.3 to 905 K and 430 to 852 K respectively, using a high temperature Calvet calorimeter. The enthalpy increments, (H⁰_T−H⁰₂₉₈), in J/mol for Cs₄U₅O₁₇(s) and Cs₂U₂O₇(s) can be represented by, H⁰_T−H⁰_298.15 (Cs₄U₅O₁₇) kJ/mol±0.9=−188.221+0.518T (K)+0.433×10⁻³T² (K)−2.052×10⁻⁵T³ (K) (368 to 905 K) and H⁰_T−H⁰_298.15 (Cs₂U₂O₇) kJ/mol±0.5=−164.210+0.390T (K)+0.104×10⁻⁴T² (K)+0.140×10⁵(1/T (K)) (411 to 860 K). The thermal properties of Cs₄U₅O₁₇(s) and Cs₂U₂O₇(s) were derived from the experimental values. The enthalpy of formation of (Cs₄U₅O₁₇, s) at 298.15 K was calculated by the second law method and is: Δ_fH⁰_298.15=−7645.0±4.2 kJ/mol.     

Thermochemical study of vaporization of LiTiO by a mass spectrometric Knudsen effusion method

The vaporization of Li₄TiO₄ has been studied by a mass spectrometric Knudsen effusion method in the temperature range 1082–1582 K. Identified vapors are Li(g), LiO(g), Li₂O(g) and Li₃O(g). When the vaporization proceeds, the content of Li₂O in the Li₄TiO₄ sample decreases and the condensed phase of the sample changes to β-Li₄TiO₄ plus l-Li₂TiO₃ below 1323 K, to β-Li₄TiO₄ plus h-Li₂TiO₃ in the range 1323–1473 K and to h-Li₂TiO₃ plus liquid above 1473 K. On the basis of the partial pressure data, the enthalpies of formation for β-Li₄TiO₄ from elements and from constituent oxides have been determined to be ΔH_f,298^°(β-Li₄TiO₄,s) = −2247.8 ± 14.3 kJ mol⁻¹ and Δ_fox,298^°(β-Li₄TiO₄, s) = −107.3 ± 14.3 kJ mol⁻¹, respectively.     

The thermal conductivity of UO2 vapor

The thermal conductivity, λ of a saturated vapor over UO_1.96 is calculated in the temperature range 3000–6000 K. The calculation shows that the contribution to λ from the transport of reaction enthalpy dominates all other contributions. All possible reactions of the gaseous species UO₃, UO₂, UO, U, O, and O₂ are included in the calculation. We fit the total thermal conductivity to the empirical equation λ = exp(a+ b/T+cT+dT² + eT³), with λ in cal/(cm s K), T in kelvins, a = 268.90, B = − 3.1919 × 10⁵, C = −8.9673 × 10⁻², d = 1.2861 × 10⁻⁵, and E = −6.7917 × 10⁻¹⁰.     

Enthalpy and heat capacity of LiAlO2 between 298 and 1700 K by drop calorimetry

The enthalpy of γ-LiAlO₂ was measured between 403 and 1673 K by isothermal drop calorimetry. The smoothed enthalpy curve between 298 and 1700 K results in H⁰(T) − H⁰(298 K)=−37 396 + 93.143 · T + 0.00557 · T² + 2 725 221 · T⁻¹ J/mol. The standard deviation is 2.2%. The heat capacity was derived by differentiation of the enthalpy curve. The value extrapolated to 298 K is C_p,298=(65.8 ± 2.0) J/K mol.     

Phase transitions in U3O8−z: I, heat capacity measurements

The heat capacity of U₃O_8−z with various O/U ratios was measured in the range from 250 to 750 K, and λ-type heat capacity anomalies were found in each sample. The transition temperatures were 487 and 573 K for UO_2.663, 490 and 576 K for UO_2.656 and 508, 562 and 618 K for UO_2.640. The entropy changes of the transitions were 0.44 and 0.39 J K⁻¹mol⁻¹ for UO_2.663, 0.58 and 0.47 J K⁻¹mol⁻¹ for UO_2.656 and 0.62, 0.51 and 0.25 J K⁻¹mol⁻¹ for UO_2.640, increasing as O/U decreases. The enthalpy change due to the transition varied linearly with the transition temperature except for UO_2.640, showing the presence of the same mechanism of phase transition among the samples with various O/U ratios. The mechanism of the phase transition was discussed on the assumption that the transition is originated from the order-disorder rearrangement of U⁵⁺ and U⁶⁺ with a consequent displacement of atoms, similarly to the case of U₄O_9−y.     

The novel ultrastable HVEE 3.5 MV Singletron accelerator for nanoprobe applications

Recently, HVEE has completed a novel 3.5 MV single ended accelerator (Singletron^™) for the University of Leipzig, Germany. For one of the main applications, the system will be connected to a nanobeamline to achieve submicron resolution. Because the energy stability and ripple of the beam, and beam brightness are of vital importance for the performance of a nanoprobe, special care has been taken in optimizing these parameters. The system consists of an RF source which is directly mounted on the accelerator tube, a switching magnet to bend the beam into a chamber for standard analysis purposes and an analysis magnet that directs the beam into the nanoprobe. The stability of the beam energy was measured at a terminal voltage of 1.881 MV. These measurements were taken during factory acceptance with large production equipment operational, which negatively influenced the stability of the mains. The measured stability was found to be approx. ±50 eV over 5 h, but it is anticipated that this figure will be as good as ±20 eV (i.e. 10⁻⁵) under normal laboratory conditions. The terminal voltage ripple was measured at 2.25 MV to be 25 V_pp (i.e. 1.1 × 10⁻⁵). Finally, the beam brightness of a 2.25 MeV hydrogen beam was measured by the use of two micrometer slit systems. A brightness of approx. 18 Amps · rad⁻² m⁻² eV⁻¹ was obtained. In this article we will describe the considerations which have led to the layout of the present system.     

Effect of tin on the irradiation growth of polycrystalline zirconium

Measurements of irradiation growth of polycrystalline Zr-1.5% Sn and Zr-0.1% Sn alloys at 353 K and 553 K have been made following fast neutron irradiation with fluences up to 3.1 × 10²⁵ n/m². At 353 K, growth of Zr-1.5% Sn virtually saturated at a strain of 4.5 × 10⁻⁴ after a fluence of ˜10²⁴ n/m². At this temperature, Zr-0.1% Sn continued to grów until ˜ 2 × 10²⁵ n/m², when the strain levelled off at ˜ 1.2×10⁻³. At 553 K, Zr-1.5% Sn initially grew about twice as fast as the 0.1% Sn alloy, but both eventually reached the same steady state rate of ˜ 2.4 × 10⁻²⁹ m²/n. Comparison of the data for the 1.5% Sn material with those for Zircaloy-2 from earlier work reveals that at 353 K, growth is suppressed by the presence of Sn atoms, which may serve as vacancy traps. However, at 553 K, minor additions and impurities in Zircaloy-2 (such as Fe, Ni, Cr and O) play an important role and cannot be neglected. The growth behaviour of Zr-0.1% Sn is similar to that of pure polycrystalline zirconium, especially at 353 K, indicating that the addition of Sn at this concentration does not strongly influence the growth of zirconium.     

Defect evolution during ion bombardment of amorphous Si probed by in situ conductivity measurements

In this work we use in-situ conductivity measurements during ion irradiation as a sensitive probe of the defect structure of amorphous Si. Electronic transport in amorphous Si occurs by hopping at the high density ( 10²⁰ cm⁻³ eV⁻¹) of deep lying localized states introduced by the defects in the band gap. In-situ conductivity measurements allow to follow directly the defect generation and annihilation kinetics during and after ion bombardment of the material. Amorphous Si layers, patterned to perform conductivity measurements, were annealed at 500°C in order to reduce the defect density by about a factor of 5. Defects were subsequently reintroduced by high energy ion irradiation at different temperatures (77–300 K). During irradiation the conductivity of the layer increases by several orders of magnitude and eventually saturates. Turning off the beam results in a decrease of the conductivity by a factor of 2 in times as long as a few hours even at 77 K. The effects of different ions (He, C, Si, Cu, and Au) and different ion fluxes (10⁹–10¹² ions/cm² s) on these phenomena have been explored. These data give a hint on the mechanisms of defect production and annihilation and demonstrate a strong correlation between electrical and structural defects in amorphous silicon.     

热压烧结UN陶瓷芯块的性能

UN燃料具有铀密度高、熔点高、热导率高、热膨胀系数低、辐照稳定性好等优点,是未来空间核电源、核火箭、快堆和ADS的重要候选燃料。本文采用金属铀粉与氮气在300～400℃直接发生化合反应,制得单相U₂N₃粉末。粒度为38.3 μm的U₂N₃粉末在1 600 ℃真空热压烧结,制得相对密度为93.5%、存在少量金属铀相的UN陶瓷;而18.1 μm的U₂N₃粉末在1 550 ℃真空热压烧结,制得相对密度为96.1%、不残留金属铀相的UN陶瓷,U与N的总质量分数为99.57%,每个金属杂质含量均低于50 μg/g,氧含量为1 048 μg/g,碳含量为502 μg/g。U₂N₃在1 027 ℃以上将会完全分解成UN,UN在1 627 ℃以上也会发生分解。